Iron Phosphate Glass as Optional Final Waste Matrix for High-Level Radioactive Waste

2003 ◽  
Author(s):  
M. Sazarashi ◽  
K. Maruyama ◽  
S. Ono ◽  
K. Suzuki ◽  
T. Fukui ◽  
...  
Keyword(s):  
Radioactive Waste ◽  
Phosphate Glass ◽  
Glassy State ◽  
Iron Phosphate ◽  
Leaching Rate ◽  
Spent Fuel ◽  
Phosphorus Oxide ◽  
High Level Radioactive Waste ◽  
P Matrix ◽  
High Level

The study investigated experimentally a basic capability of iron phosphate glass (IPG) matrix to vitrify the high level radioactive waste (HLW) in the commercial reprocessing. IPG matrix is a mixture of iron oxide and phosphorus oxide, and Fe/P mole ratio is one of important conditions to form a vitrified product with good chemical and physical properties. Moreover, the corrosion of IPG matrix is more rigorous than that of borosilicate glass matrix, and an appropriate furnace material should be selected for the durability of equipments. The composition of HLW used in experiments was simulated, based on that of HLW generating from commercial reprocessing for a standard spent fuel of light water reactor. Dependence of Fe/P mole ratio on both crystallinity of product and loading of simulated HLW was investigated. At Fe/P mole ratios of 0.43 and 0.33, crystals of Nd and Zr phosphate, appeared in a part of Fe-P matrix zone of vitrified products containing 30wt% of the HLW. The products at Fe/P ratios of 0.18 and 0.27 become a homogeneous glassy state with increasing solubility of crystals of Nd and Zr phosphate. It means that the low Fe/P mole ratio is better to form the vitrified product of high waste loading. Concerned with water resisting property, products vitrified at Fe/P = 0.18,0.23 have larger amount of dissolution rates than that at Fe/P = 0.33,0.43. An appropriate amount of Fe needs to form vitrified product of IPG with good water resisting property, although the homogeneity in micro-structure of the product slightly decreases. Fe/P mole ratio in IPG, 0.3 is an appropriate condition to form vitrified product containing HLW with less dissolution and leaching rate. It is totally concluded that Fe/P mole ratio in IPG, 0.3 is an appropriate condition to form vitrified product containing HLW with less dissolution and leaching rate. This study focused on a special material, ZrB2, which has electric and heat conduction equivalent to iron metal in order to estimate an adoption of furnace material. The corrosion rate of ZrB2 in the melting of IPG matrix was as low as that of Alumina, and there was no rigorous corrosion on the surface of crucible. It was shown that ZrB2 is able to apply a furnace material for IPG vitrification with an induction-heated ceramic melter.

A Review of Corrosion Considerations of Container Materials Relevant to Underground Disposal of High-Level Radioactive Waste in Belgium

2006 ◽  
Vol 932 ◽  
Author(s):  
Bruno Kursten ◽  
Frank Druyts
Keyword(s):  
Radioactive Waste ◽  
Large Scale ◽  
General Corrosion ◽  
Spent Fuel ◽  
High Level Radioactive Waste ◽  
Boom Clay ◽  
Nuclear Site ◽  
In Situ Experiments ◽  
Underground Disposal ◽  
High Level

ABSTRACTThe underground formation that is currently being considered in Belgium for the permanent disposal of high-level radioactive waste and spent fuel is a 30-million-year-old argillaceous sediment (Boom Clay layer). This layer is located in the northeast of Belgium and extending under the Mol-Dessel nuclear site at a depth between 180 and 280 meter.Within the concept for geological disposal (multibarrier system), the metallic container is the primary engineered barrier. Its main goal is to contain the radioactive waste and to prevent the groundwater from coming into contact with the wasteform by acting as a tight barrier. The corrosion resistance of container materials is an important aspect in ensuring the tightness of the metallic container and therefore plays an important role in the safe disposal of HLW. The metallic container has to provide a high integrity, i.e. no through-the-wall corrosion should occur, at least for the duration of the thermal phase (500 years for vitrified HLW and 2000 years for spent fuel).An extensive corrosion evaluation programme, sponsored by the national authorities and the European Commission, was started in Belgium in the mid 1980's. The main objective was to evaluate the long-term corrosion performance of a broad range of candidate container materials. In addition, the influence of several parameters, such as temperature, oxygen content, groundwater composition (chloride, sulphate and thiosulphate), γ-radiation, … were investigated. The experimental approach consisted of in situ experiments (performed in the underground research facility, HADES), electrochemical experiments, immersion experiments and large scale demonstration tests (OPHELIE, PRACLAY). Degradation modes considered included general corrosion, localised corrosion (pitting) and stress corrosion cracking.This paper gives an overview of the more relevant experimental results, gathered over the past 25 years, of the Belgian programme in the field of container corrosion.

A Computational Model for Estimating the Cost of Spent Fuel and High-Level Radioactive Waste Transport

2003 ◽  
Author(s):  
E. R. Johnson ◽  
R. E. Best
Keyword(s):  
Radioactive Waste ◽  
Spent Nuclear Fuel ◽  
Cost Model ◽  
Transportation Cost ◽  
Cost Calculation ◽  
Spent Fuel ◽  
Transport Costs ◽  
High Level Radioactive Waste ◽  
High Level ◽  
The Cost

JAI has developed a simple computer program for use in determining a preliminary estimate of costs for transporting spent nuclear fuel or high-level radioactive waste by legal weight truck or by rail. The JAI Corporation Spent Fuel and High-Level Radioactive Waste Transportation Cost Model © is a Microsoft Excel 2000-based collection of spreadsheets. Both the truck and rail sub-models consist of three spreadsheets, or modules — as follows: • The “Input” spreadsheet accepts the user’s inputs (the user’s configuration of the transportation scenario to be modeled); • The “Cost Calculations” spreadsheet lists cost components and associated calculations; • The “Results” spreadsheet summarized the calculated transportation costs. The program does not calculate costs between two specific points, but rather over a specific distance. The individual inputs required can be entered by the user — or the user can accept the default values built into the program. The input to the program is divided into the following elements: 1. Scenario configuration; 2. Financial assumptions; 3. Capital-related costs; 4. Operating costs; 5. Freight-related costs; 6. Security-related costs. The rail portion of the program also permits the calculation of the cost of heavy haul and barge transport. The cost calculation spreadsheet contains all the algorithms used for calculating each element of cost and summing them — and the results spreadsheet shows the separate cost of capital, operations, freight, security and miscellaneous costs, plus the total cost for the shipment(s). The program offers an easy way for obtaining preliminary estimates of the cost of transporting spent fuel or high-level radioactive waste, and a way to quickly estimate the sensitivity of transport costs to changes in conditions or shipping scenarios.

Localized Corrosion and Stress Corrosion Cracking of Candidate Materials for High-Level Radioactive Waste Disposal Containers in U.S.: A Critical Literature Review

1988 ◽  
Vol 127 ◽  
Author(s):  
Joseph C. Farmer ◽  
R. Daniel McCright
Keyword(s):  
Radioactive Waste ◽  
The United States ◽  
Radioactive Waste Disposal ◽  
Localized Corrosion ◽  
Spent Fuel ◽  
High Level Radioactive Waste ◽  
Fuel Assemblies ◽  
Austenitic Alloys ◽  
High Level ◽  
Critical Literature

ABSTRACTThree iron-based to nickel-based austenitic alloys and three copper-based alloys are being considered in the United States of America as candidate materials for the fabrication of high-level radioactive waste containers. The austenitic alloys are Types 304L and 316L stainless steels as well as the high-nickel material Alloy 825. The copper-based alloys are CDA 102 (oxygen-free copper), CDA 613 (Cu-7A1), and CDA 715 (Cu-3ONi). Waste in the forms of spent fuel assemblies from reactors and borosilicate glass will be sent to a proposed repository at Yucca Mountain, Nevada. The decay of radionuclides will result in the generation of substantial heat and in gamma radiation.

Safety Assessment of Geological Disposal of High-Level Radioactive Waste in Boom Clay: Relation With the Radionuclide Inventory

2009 ◽  
Author(s):  
Pierre Van Iseghem ◽  
Jan Marivoet
Keyword(s):  
Radioactive Waste ◽  
Safety Assessment ◽  
High Level Waste ◽  
Spent Fuel ◽  
Geological Disposal ◽  
High Level Radioactive Waste ◽  
Boom Clay ◽  
High Level ◽  
Clay Formation ◽  
Parameter Values

This paper discusses the impact of the parameter values used for the transport of radionuclides from high-level radioactive waste to the far-field on the long-term safety of a proposed geological disposal in the Boom Clay formation in Belgium. The methodology of the Safety Assessment is explained, and the results of the Safety Assessment for vitrified high-level waste and spent fuel are presented. The radionuclides having the strongest impact on the dose-to-man for both HLW glass and spent fuel are 79Se, 129I, 126Sn, 36Cl, and 99Tc. Some of them are volatile during the vitrification process, other radionuclides are activation products, and for many of them there is no accurate information on their inventory in the waste form. The hypotheses in the selection of the main parameter values are further discussed, together with the status of the R&D on one of the main dose contributing radionuclides (79Se).

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